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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Assessment of operational availability for the PIP-II Superconducting Radio Frequency linear accelerator facility

Operational availability is a critical performance measure for an accelerator facility in modern time. A high availability enables the facility to serve a wide range of users simultaneously. Consequently, besides pure accelerator physics considerations, newly proposed accelerator facilities account for the availability and reliability aspects in the design phases. It allows incorporation of appropriate mitigation strategies for the most vulnerable systems in the machine and therefore, minimizes unscheduled interruptions during the operation. This report lays out a methodology for the availability assessment of the complete particle accelerator facility and presents an initial assessment of the availability of the newly proposed Proton Improvement Plan-II (PIP-II) accelerator facility at Fermilab. The paper describes a comprehensive reliability model of the PIP-II facility that comprises not only 800 MeV linear accelerator (linac) system but also essential utility systems in the form of cryogenic, water, power and air systems. The paper details estimations of the availability of the PIP-II facility for two operational modes i.e. the nominal operational mode featuring 800 MeV beam and critical operational mode involving operation with the lowest objective beam energy of 600 MeV.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Passive Safety Features Evaluation of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. It will be used to conduct basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The facility has an electron accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 - Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The subcritical assembly is seated in a water tank, which is a part of the primary cooling loop. During normal operation, the water coolant operates at room temperature and the total facility power is ~300 KW. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B-VII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight water temperature increase and the corresponding water density decrease produce large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. The reactivity variation due to an empty fuel position filled by water during the fuel loading process is examined. Also, the loading mistakes of removing beryllium reflector assemblies and replacing them with dummy assemblies were analyzed. In all these circumstances, the reactivity change results do not cause any safety concerns.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Dose Modeling for Harborview Research and Training Facility

The Harborview Research and Training (R&T) Facility at the University of Washington (UW) was accidently contaminated from a breached sealed source of Cesium-137. The release occurred during removal of the sealed source from its holder within a portable hot cell in the loading dock area near room 220. The loading dock and northeast corner of the building above the ground floor were impacted by the release, whereas the remaining portions of the R&T Facility were minimally affected. Detailed descriptions of the R&T Facility and the affected area/rooms are included in Section 2. The RESRAD Team in the Environmental Science Division of Argonne National Laboratory visited the site on August 12, 2019, and interviewed the R&T Facility building manager and staff working in the building at that time. Site- and building-specific data were collected including building floor plan CAD drawings, building occupants’ working hours, contaminated rooms photos and their dimensions, building air exchange rates, building materials and “curtain wall” details, and so forth. These data are being used in the dose modeling for potential occupants of the building after it is remediated. The dose modeling was performed using the RESRAD-BUILD computer code, which is designed for the evaluation of radiologically contaminated buildings (Yu et al. 2003). Two hypothetical building occupants of the Harborview R&T Facility were evaluated: a researcher occupying room 726 (a research laboratory on the seventh floor) and a custodian occupying room 220 (loading dock and trash compactor room with a rollup door). A maintenance worker that will be spending less time in impacted areas was not modeled because it will be bounded by other two hypothetical building occupants modeled. A summary description of the RESRAD-BUILD is provided in Appendix A, and the modeling of potential doses to a researcher and a custodian is described in detail in Section 3. A summary and a discussion of the dose modeling results are included in Section 4. The parameters used in the RESRAD-BUILD modeling are included in Appendix B. This report documents the dose modeling results; it does not describe measurements of the actual radioactivity levels at the Harborview R&T Facility, nor does it set any expectation for the actual final radioactivity levels. The radiation doses presented are modeling results based on the generic screening values of Cs-137 and several conservative assumptions and are meant only to assist in decision making regarding decontamination of the Harborview R&T Facility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Public Views on Nuclear Facility Siting and Radioactive Waste Management in the United States: Methodology and Response Reference Report for the 2017 Energy and Environment Survey

This report presents the questions and responses to a nationwide survey of U.S. residents taken May 2017. The focus of the 2017 survey is public perceptions on (1) nuclear energy and new nuclear technology, such as small modular reactors; (2) the risks and benefits of nuclear energy; (3) different options for managing spent nuclear fuel (SNF); and (4) the role of individuals, organizations and elected officials in locating SNF facilities. Highlights of survey results related to nuclear energy found the U.S. public (1) wants nuclear energy to continue to contribute 20% of total electrical power produced for the next 20 years; (2) is more supportive of adding new reactors at existing sites than at new sites; and (3) is more supportive of small modular reactors than traditional reactors. Highlights related to risks and benefits of nuclear energy include (1) the public believes that terrorist attacks are the greatest threat and that energy independence is the greatest benefit; (2) 56% of the public believes that nuclear power has both high benefit and high risk; (3) the 2011 accident at the Fukushima nuclear facility continues to negatively influence the perceived benefits and risks of nuclear energy except for those who perceive nuclear power as low risk and high benefit; (4) building new reactors is supported by those who perceive nuclear power as low risk and high benefit, and opposed by those who perceive it as high risk and low benefit. Highlights related to SNF management options include (1) support remains higher for permanent disposal than either consolidated interim storage or continued on-site storage; (2) support for private initiatives for interim storage in New Mexico and Texas are higher than generic storage and disposal options; (3) the general public is more likely to increase support if a majority of the local and state residents support an SNF facility. Highlights related to the role of individuals, organizations, and elected officials include (1) 54% of the public agrees that if an SNF facility is deemed safe by regulators, political considerations should not derail the location; (2) 47% of the US public agrees that non-governmental organizations should be able to provide input to locating an SNF facility; (3) 36% of the public believes that the federal government should lead siting of an SNF facility while 25% believes private companies should lead; (4) 81% of the public believes local residents and farmers/ranchers should have a say in the final decision on locating a SNF facility and 56% believe that they should have veto power; and (5) only 27% of the public believes the state governor should have a say and veto power.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

E-Area Low-Level Waste Facility Multitiered Groundwater and Intruder Radionuclide Screening

Solid Waste Management (SWM) operates the E-Area Low-Level Waste Facility (ELLWF) where low level on-site and off-site solid waste streams are buried. The facility has been in operation since late 1994 and is currently projected to remain in operation until 2065. This facility can accommodate a broad range of waste forms resulting from the six different types of disposal unit (DU) options (i.e., varying degrees of engineered barriers → trenches to concrete vaults). This facility is currently operating under a Performance Assessment (PA) issued back in 2008 (WSRC 2008), along with several subsequent supporting Special Analyses (SAs). The Savannah River National Laboratory (SRNL) developed the prior PAs and SAs and was tasked to update the facility’s upcoming PA, most likely to be issued during FY2023. For operating the E-Area facility, a Consolidated Waste Tracking System (CWTS) is actively employed by waste generators where every radionuclide entering the facility, to be buried in one of its many DUs1, must be either directly or indirectly tracked. Since there are many radionuclides in existence (>3,000), the International Commission on Radiological Protection (ICRP), specially ICRP Publication 107, has provided guidance on the subset of radionuclides requiring further assessment in landfills such as the ELLWF. The ICRP 107 publication provides critical radiological information on 1,252 radionuclides consisting of 97 elements. This database, along with the current dose coefficients that have been developed for these radionuclides in DOE-STD-1196-2011 (DOE 2011), is the critical starting point for developing a consistent inventory limit system.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Baseline Hypothetical Facility for the Production of 131 I and 99 Mo using Activation Targets

This report describes a hypothetical facility for production of medical radioisotopes via activation under the Proliferation Resistance and Optimization (PRO-X) program. The facility uses neutron activation of non-special nuclear material (SNM) to produce the medical isotopes 131 I and 99 Mo at a throughput of 60 Ci/week of 131 I and 5 Ci/week of 99 Mo. The hypothetical design was carried out using a 10 MWt research reactor. The precursors used for the activation process were TeO2 for 131 I and MoO 3 for 99 Mo. The processes are performed in 3 hot cells used for target receipt, extraction, purification low specific activity (LSA) generator introduction, and packaging. A fourth hotcell is used for waste processing. The hot cell processing area takes up a footprint of 15.4 m 2 with the total footprint of the facility, including space for administrative offices, non-rad labs, quality assurance, and radiation buffer areas set at 763 m 2 . Waste is produced at a weekly rate of 257.8 g low activity solid waste and 8032.7 mL of low activity liquid waste, 8032 mL of which is water. This baseline hypothetical facility for production of medical isotopes via activation was then compared and contrasted to the hypothetical facility for production of medical isotopes via fission products to show the differences in approach for the two production modes. The two production modes had several highlighted differences including the overall facility and hot cell layout, the type and amount of waste produced by the respective facilities, and economic factors impacting production mode. Finally, a decision tree for which production mode might be more beneficial for an entrant into medical isotope production was developed based on the differences examined and the desired output of medical isotopes desired by the entrant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Recent Upgrades at the Safety and Tritium Applied Research Facility

This paper gives a brief overview of the Safety and Tritium Applied Research (STAR) facility operated by the Fusion Safety Program (FSP) at the Idaho National Laboratory (INL). FSP researchers use the STAR facility to carry out experiments in tritium permeation and retention in various fusion materials, including wall armor tile materials. FSP researchers also perform other experimentation as well to support safety assessment in fusion development. This lab, in its present two-building configuration, has been in operation for over ten years. The main experiments at STAR are briefly described. This paper discusses recent work to enhance personnel safety at the facility. The STAR facility is a Department of Energy less than hazard category 3 facility; the personnel safety approach calls for ventilation and tritium monitoring for radiation protection. The tritium areas of STAR have about 4 to 12 air changes per hour, with air flow being once through and then routed to the facility vent stack. Additional radiation monitoring has been installed to read the laboratory room air where experiments with tritium are conducted. These ion chambers and bubblers are used to verify that no significant tritium concentrations are present in the experiment rooms. Standby electrical power has been added to the facility exhaust blower so that proper ventilation will now operate during commercial power outages as well as the real-time tritium air monitors.

tritium facility upgrades↗
DOE OSTI2018

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory designed to perform transient testing on nuclear fuel experiments and materials. A neutron radiography facility was installed with the first radiograph being produced in 1967 that utilized the indirect (or transfer) method of neutron radiography to examine irradiated nuclear fuels and experiment test loops. The TREAT neutron radiography facility performed approximately 5,000 radiographs by the spring of 1977. The original radiography facility was upgraded in 1974-1975 to correct previous deficiencies and to improve capabilities. In 1986, an upgraded aperture was installed to increase the resolution of images followed by a major reactor upgrade completed in 1988. Overall, the existing neutron radiography system was in good working order. Originally built in 1958, the TREAT Facility was in operation until it was placed in a shutdown status in 1994. Following the Fukishima disaster and seeing a need for enhanced accident tolerant fuels, the United Stated Department of Energy decided to restart the TREAT facility and resume transient operations. In November 2017, the TREAT reactor was successfully restarted and is currently performing operational testing in preparation for initial experiment irradiations and transient testing. This paper reports on efforts to reactivate the TREAT neutron radiography station. A few minor shielding and system improvement upgrades have been completed along with the installation of a dedicated room for computed radiography equipment. Detailed beam characterization work was not performed following the last reactor upgrade. To characterize the neutron beam, gold foil activation measurements were made to determine an average neutron flux and flux profile. Additional multi-foil activation measurements will be made to calculate the energy spectrum. An open beam image will be taken that will provide the information about variations in the beam profile. A series of system qualification shots will also take place to accurately determine the effective image quality and image acquisition parameters. Full operational status of the TREAT neutron radiography system is scheduled for late summer of 2018.

36 MATERIALS SCIENCE↗
DOE OSTI2015

The Industrial Experience Feedback and R and D Ensure the Longevity of French Vitrification Facilities - 15503

The French vitrification of High Level Liquid Waste (HLLW) developed since the beginning of the 1960's led to the successive completion of four industrial vitrification facilities: PIVER facility (1969), Marcoule vitrification facility (1978), and R7 (1989) and T7 (1992) vitrification facilities on La Hague site. R and D has involved containment glass formulation and characterization of glass properties, development of modular technologies specifically tailored to be operated in shielded cells, and development of the vitrification process on full scale prototype. Industrial support has then provided by the R and D team during all phases of commissioning and operation of these industrial facilities. The initial choices of industrial facility design, as well as technologies to enhance flexibility and modularity, allowed an evolution of the vitrification process over time. Maintaining the R and D capabilities on containment glass, process and technology development at full scale size is essential to anticipate treatment of the new effluent and propose innovative industrial technologies. Fueling R and D with industrial feedback ensured the longevity of industrial vitrification facilities and development capabilities for future projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Progress on Resuming Full Construction on the Waste Treatment and Immobilization Plant's High-Level Waste Facility - 16283

The U.S. Department of Energy (DOE) restricted certain engineering, procurement and construction work on the Waste Treatment and Immobilization Plant's (WTP) High-Level Waste (HLW) Facility in 2012 and early 2013 because of unresolved technical issues, and a need to align the nuclear safety basis with the design as the HLW Facility technical issues were resolved. The DOE Office of River Protection (ORP) needed a way to minimize the government's risk exposure, while continuing limited critical path construction work and resolving the technical and safety basis issues. ORP issued the Plan for Evaluation and Decision to Proceed for the High-Level Waste Facility in October 2013 to establish the process to evaluate the WTP contractor's progress on resolving the issues and to identify actions and deliverables needed to facilitate an ORP decision to resume production activities. This process provided for a phased resumption of HLW engineering, procurement and construction activities, with two decision points: Decision 1: Authorization to resume design/production engineering along with limited procurements and construction ramp-up. Decision 2: Full authorization to resume engineering, procurement and construction. In August 2014, ORP provided authorization for the WTP contractor to resume all engineering work necessary to finalize the design of the HLW Facility. The decision to proceed with production engineering in the HLW Facility was based on significant progress in resolving technical issues and improving work processes to align the HLW design and safety basis. ORP is working closely with the WTP contractor to facilitate decisions for resuming procurement of HLW Facility equipment with safety functions, and resuming full construction in the most complex process cells in the Facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Decommissioning, Decontamination, and Disposal of LLW-LL, and ILW-LL at a Former Rare Earths Processing Facility in Malaysia - 19087

Geosyntec Consultants, Inc. (Geosyntec) was the design-build Contractor for a project involving the remediation and disposal of low level (LLW-LL and ILW-LL) NORM and TENORM wastes at a former rare earth processing facility in Malaysia. These sites were associated with a plant that extracted rare earth elements from local tin ores. Because of the naturally occurring radioactivity contained in the raw materials, these productions resulted in by-products and residues containing LLW-LL and ILW-LL thorium hydroxide and uranium hydroxide. The total mass of uranium (∼120 tons) and thorium (∼1,200 tons) managed in a single repository required the project design incorporate requirements related to inventory, security and safeguards around nuclear precursor materials. The requirement to consider a performance over a period of more than 10,000 years led to significant challenges in design and modeling of the disposal facility as this disposal facility was purposely designed and permitted as a final disposal for the LLW-LL and ILW-LL. The first phase of the project was the decommissioning, decontamination, and demolition of the 14- acre former rare earth processing plant and excavation of contaminated soils, construction of an engineered cell at the repository for waste disposal, and transportation and disposal of the waste materials into the cell. Final status surveys were carried out to allow the site to be successfully released by the Malaysian regulatory authorities as a clean site for re-development as common industrial use. Over 11,000 loads of material, including approximately 75,000 cubic meters of excavated soils and plant debris, were transported in specially designed steel transport boxes on public roads for disposal in the engineered cell. The second phase of the project included the design and construction of a second engineered cell, enhanced with protective barrier systems and a final cover system constructed of natural materials. Ancillary facilities such as substations, offices, processing buildings, and wastewater treatment plants were constructed to support retrieval and processing of the materials from more than 87,000 individual drums using a semi-automated process. The drums, waste packages, and other materials in the storage facility were systematically extracted, sorted, inventoried, solidified using Portland cement, transported, and disposed in the cell. All temporary facilities were decontaminated, demolished, and disposed in the cell. Radiological inventory and radiation safety programs were implemented to quantify the waste activity and to control exposure to workers and the public during the project. The total mass of radioactive contaminated waste disposed in the two projects is nearly 200,000 tons. The projects were carried out under a high level of scrutiny from Malaysian and international regulatory bodies. Between 2014 and 2017, performance monitoring and modeling activities were carried out to demonstrate that the disposal facility was performing in accordance with design requirements established by regulatory agencies. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Development of a Small-Scale Project Model for Waste Management at the Port Hope Conversion Facility - 19126

Cameco Corporation's (Cameco) Port Hope Conversion Facility (PHCF) was previously owned by the federal Crown Corporation, Eldorado Nuclear, which held a significant inventory of legacy waste material at the time Cameco was formed in 1988. Over time, Cameco's ongoing operations contributed to an increase in the volume of low-level radioactive waste managed onsite. Since 2011, Cameco has completed annual waste management campaigns with the goal of accelerating the recycle/disposal of accumulated legacy radioactive wastes stored at its Ontario operations. With disposal options for low-level radioactively contaminated wastes in Canada limited due to cost considerations and no commercially accessible landfills licensed for radioactive material. In recent years, Cameco has started to ship marginally contaminated wastes (<500 ppm uranium) to the United States for disposal in an appropriate landfill owned by US Ecology in Idaho. Numerous successful campaigns totaling over 1,000 successful shipments, have been completed to transport drummed and/or bagged material by rail in gondola cars to the disposal facility in Idaho since 2011. All shipments have been completed in compliance with all environmental regulations, with a perfect compliance record maintained to date. To achieve this excellent compliance record, the campaigns require careful planning by both organizations, as well as deployment of personnel from the Idaho facility during the campaign for project management, characterization and/or radiation monitoring support. As the accumulated waste inventory has been reduced through these projects, the scope and duration of each campaign has also decreased. In the last few years, road transport directly to the Idaho facility has also been utilized. This option allows for the deployment of small projects that may be completed for a smaller inventory of a specific waste type and to manage newly-generated low-level radioactive wastes on an ongoing basis. These projects have ranged from a single standard van shipment to multiple trucks running continuous loops between the Ontario and Idaho or Ontario and an intermediary site and have been utilized for marginally contaminated wastes (<500 ppm uranium) as well as marginally contaminated wastes with additional hazard classification. Cameco and US Ecology personnel have collaborated to develop characterization and shipment procedures that enable these projects to be executed in full compliance with all Canadian and US regulations, without deployment of personnel from the Idaho facility, while maintaining adequate oversight by both organizations to ensure that the waste is safely and compliantly managed. These protocols have been implemented and optimized for the Port Hope Conversion Facility and will be rolled out to other Cameco facilities in Ontario through 2019. Cameco expects to utilize the small project model for the long term to ensure that eligible marginally contaminated wastes (<500 ppm uranium) are managed appropriately to ensure that once the legacy material is eliminated, it remains so. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Saltstone Facility Operating Experience with Mega-Vault Disposal Units - 19497

The Saltstone Production Facility (SPF) has been treating low-level waste aqueous salt solution at the Savannah River Site (SRS) since 1990. Salt solutions are transferred from a large-volume, underground storage tank via underground waste transfer line into the SPF. The waste is received into one of two receipt tanks and then transferred to a twin-shaft mixer where dry feeds are introduced to create a fresh grout slurry. The grout exits the mixer into a hold tank, which has a residence time of approximately two minutes. The slurry is then pumped via peristaltic pump through a grout transfer line to one of several disposal units in the Saltstone Disposal Facility (SDF). The transfer line length varies from 0.4 km to 0.8 km depending on disposal location. With the salt disposal mission at SRS accelerating via the operation of the Modular Caustic Side Solvent Extraction (MCU) facility and the anticipated startup of the Salt Waste Processing Facility (SWPF), it is important that the SPF and SDF operate with minimal outage time and maximum facility availability. To this end, the design of the newest disposal units has incorporated the benefits of commercially available cylindrical concrete tank design on a very large scale to maximize economies of scale. This new design has been placed into radioactive service and the initial operational data will be compared with previous disposal operations along with an overview of how disposal unit design has significantly evolved since the initial construction in the early 1990's. Additionally, several processing facility modifications have either been implemented or developed in recent years to improve operating reliability and ensure facility availability once the SWPF becomes the primary waste generator providing feed for the SPF. These modifications, the issues driving their development, and an assessment of their effectiveness will also be discussed. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Passive Safety Feature of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. The facility has an electron accelerator driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The water coolant operates at room temperature during normal operation, when the facility is operating at full power using 100-KW electron beam power. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/BVII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight temperature increase and the corresponding density decrease in the water coolant would cause relatively large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Addressing social aspects associated with wastewater treatment facilities

In wastewater treatment facilities (WWTF), technical and financial aspects have been considered a priority, while other issues, such as social aspects, have not been evaluated seriously and there is not an accepted methodology for assessing it. In this work, a methodology focused on social concerns related to WWTF is presented. The methodology proposes the use of 25 indicators as a framework for measuring social performance to evaluate the progress in moving towards sustainability. The methodology was applied to test its applicability and effectiveness in two WWTF in Mexico (urban and rural). This evaluation helped define the key elements, stakeholders and barriers in the facilities. In this context, the urban facility showed a better overall performance, a result that may be explained mainly by the better socioeconomic context of the urban municipality. Finally, the evaluation of social aspects using the semi-qualitative approach proposed in this work allows for a comparison between different facilities and for the identification of strengths and weakness, and it provides an alternative tool for achieving and improving wastewater management. - Highlights: • The methodology proposes 25 indicators as a framework for measuring social performance in wastewater treatment facilities. • The evaluation helped to define the key elements, stakeholders and barriers in the wastewater treatment facilities. • The evaluation of social aspects allows the identification of strengths and weakness for improving wastewater management. • It provides a social profile of the facility that highlights the best and worst performances.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Quantification of greenhouse gas emissions from a biological waste treatment facility

Highlights: • Whole-site fugitive emissions were measured at a biowaste treatment facility. • Methane and nitrous oxide were quantified using a tracer gas dispersion technique. •The daily average whole-site methane emission was 25.6 kg CH{sub 4} h{sup −1}. • Whole-site fugitive emissions were up to eight times higher than on-site emissions. • The emissions had a significant influence on the overall environmental impact - Abstract: Whole-site emissions of methane and nitrous oxide, from a combined dry anaerobic digestion and composting facility treating biowaste, were quantified using a tracer dispersion technique that combines a controlled tracer gas release from the treatment facility with time-resolved concentration measurements downwind of the facility. Emission measurements were conducted over a period of three days, and in total, 80 plume traverses were obtained. On-site screening showed that important processes resulting in methane emissions were aerobic composting reactors, anaerobic digester reactors, composting windrows and the site’s biofilter. Average whole-site methane emissions measured during the three days were 27.5 ± 7.4, 28.5 ± 6.1 and 30.1 ± 11.4 kg CH{sub 4} h{sup −1}, respectively. Turning the windrows resulted in an increase in methane emission from about 26.3–35.9 kg CH{sub 4} h{sup −1}. Lower emissions (21.5 kg CH{sub 4} h{sup −1}) were measured after work hours ended, in comparison to emissions measured during the facility’s opening hours (30.2 kg CH{sub 4} h{sup −1}). Nitrous oxide emission was too small for a downwind quantification. Direct on-site measurements, however, suggested that the main part of the emitted nitrous oxide came from the biofilter (about 1.4 kg N{sub 2}O h{sup −1}). Whole-site emissions were compared to emissions previously measured at different point sources on-site. Whole-site fugitive emissions were three to eight times higher than the sum of emissions measured at on-site sources. The magnitude of the emissions had a significant influence on the overall environmental impact of the treatment facility, assessed by consequential life cycle assessment. Including the higher whole-site fugitive emissions led to an increase in global warming potential, from a saving of 97 kg CO{sub 2}-eq. tonne{sup −1} of treated waste (wet weight) to a loading of 71 kg CO{sub 2}-eq. tonne{sup −1}, ultimately flipping the environmental profile of the treatment facility.

09 BIOMASS FUELS↗
DOE OSTI2022

Zero Liquid Discharge and Water Reuse in Recirculating Cooling Towers at Power Facilities: Review and Case Study Analysis

Zero liquid discharge (ZLD) systems installed at power facilities with the primary purpose of meeting water discharge regulations have the added benefit of providing high quality effluent that can be reused in the facility. This paper provides a review of water use in power sector recirculating cooling towers and a baseline assessment of on-site water reuse at natural gas combined cycle (NGCC) power facilities. Two NGCC facilities with reverse-osmosis (RO) or brine-concentrator processes followed by evaporation ponds were selected as case studies; data from these facilities were used to quantify the water, energy, and cost implications of implementing conventional and emerging ZLD technologies. At one case study facility, model results show that implementation of ZLD would reduce water withdrawals by 18%, which is less than savings associated with implementation of dry cooling but comparable to current efforts to reduce water withdrawals by increasing cycles of concentration. Implementation of ZLD using high-recovery RO resulted in a doubling of the levelized cost of water (LCOW). LCOW increased more when a brine concentrator was used. For both case studies, the ZLD system using high-recovery RO required less than 0.1% of a facilitiy's annual electricity generation and the ZLD system using a brine concentrator process required less than 0.8%. Additionally, increasing the evaporation pond area to minimize required ZLD system recovery rates and reduce system electricity costs does not reduce the LCOW. Instead, the LCOW increases because less water is recovered and more water is lost to evaporation. Furthermore, if water availability decreases or water competition/cost increases, facilities may be incentivized to maximize water recovery from ZLD systems.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2020

Clusters of COVID ‐19 in long‐term care hospitals and facilities in Japan from 16 January to 9 May 2020

Aim To clarify the association of cluster number and size of coronavirus disease 2019 (COVID‐19) in long‐term care (LTC) hospitals/facilities, general medical/welfare facilities and non‐medical/welfare facilities with morbidity and mortality in 47 prefectures during 16 January to 9 May 2020 in Japan. Methods Information on COVID‐19 clusters ( n ≥2), and morbidity and mortality of COVID‐19 was collected. Results A total of 381 clusters with 3786 infected cases were collected, accounting for 23.9% of 15 852 cumulated cases on 9 May 2020. Although the cluster number (/10 7 subjects) in LTC hospitals/facilities was significantly smaller compared with those in the other two groups, the cluster size in LTC hospitals/facilities was significantly larger than that in non‐medical/welfare facilities. Cluster numbers in general medical/welfare facilities and in non‐medical/welfare facilities were significantly positively correlated with morbidity (/10 5 ), indicating relatively early identification of clusters in these facilities. Unlike in these facilities, cluster size in LTC hospitals/facilities was significantly positively correlated with morbidity, indicating that clusters in LTC hospitals/facilities were finally identified after already having grown to a large size in areas where infection was prevalent. Multivariate logistic regression analysis showed that both cluster number and cluster size only in LTC hospitals/facilities were independently associated with higher mortality (≥median 0.64/10 5 subjects) after adjustment. Conclusions Preventive efforts against COVID‐19 outbreaks even at the early phase of the epidemic are critically important in LTC hospitals/facilities, as both the larger number and size of clusters only in LTC hospitals/facilities were independently linked to higher mortality in prefectures in Japan. Geriatr Gerontol Int 2020; 20: 715–719 .

Iritani, Osamu↗